Enstrümental nötron aktivasyon analiziyle 235U/238U oranının tayini

Abstract

ÖZET Uranyum nükleer yakıt olarak dünyada geniş ölçüde kullanılmaktadır. Nükleer reaktör yakıt çevriminde, yakıt elemanlarının hazırlanmasında, reaktörlerde kullanılmış yakıtların yeniden işlenmesinde 2j5tj/238`U izotopik oranının bilinmesi gerekir. Ayrıca uranyumun tıpta da kullanılması insan sağlığı için daha zararlı izotop olan zj5U'un miktarının da bilinmesini gerektirir. Bu çalışmada bir yenilik olarak 235U/238U izotopik oranı nötron aktivasyon analiziyle yüksek rezolüsyonlu gamma spektrometresinde, daha önce kullanılmamış olan z38U'in bozunum ürünü 2j9U ve daha önce kullanılmamış olan kısa ömürlü fısyon ürünlerinin aktivitelerinden yararlanılarak bulundu. Bu amaçla hazırlanan farklı izotopik bileşimindeki uranyum örnekleri Avusturya Üniversiteleri TRIGA Mark II reaktörü hızlı ışınlama ve ölçüm sisteminde (FIMS) ve ayrıca aynı sistemdeki 6LiD dönüştürücü varlığında ışınlandı. Oluşturulan kalibrasyon eğrilerinden yararlanılarak, izotopik bileşimi bilinmeyen uranyum örneklerinde i35U/2i8U izotopik oranının çok kısa sürede elde edilebileceği gösterildi. İstanbul Teknik Üniversitesi Nükleer Enerji Enstitüsü laboratuarlarında bulunan gamma spektrometrik sayım sistemi, yazılan aktarma programıyla bilgisayara bağlandı. Bilgisayarda elde edilen gamma spektrumları, bir dönüşüm programı yazılarak çok gelişmiş bir gamma aktivite ve nötron aktivasyon analiz programı olan GANAAS ile değerlendirilebilecek formata çevrildi. GANAAS programı kullanılarak aynı laboratuarlarda bulunan çeşitli uranyum örneklerinde, yüksek rezolüsyonlu gamma spektrometrisi kullanılarak gerçekleştirilen nötron aktivasyon analizleri ve pasif gamma ışını spektrometrisi deneyleriyle de U/ U izotopik oranı belirlendi. Nötron aktivasyonu için enstitünün TRIGA Mark II reaktörü kullanıldı. XIV

DETERMINATION OF 235U / 238U RATIO WITH INSTRUMENTAL NEUTRON ACTIVATION ANALYSIS SUMMARY The fuel of ail commercial nuclear power reactors is uranium. The natural uranium is a mixture of 2j8U, 2j5U and z34U in atomic ratio of 99.274 %, 0.7205% and 0.0056% respectively. The fuel of thermal reactors is 235U, which as already mentioned, constitutes a small fraction of the isotopic composition of the natural uranium. Therefore mostly, enriched uranium is used in thermal reactors. The isotopic ratio of 35U in enriched uranium, which is used in commercial nuclear power reactors, ranges from 1.5 % to 3.5 %. Enrichment is a complicated process and isotopic analysis is needed in every step of the process in enrichment plants. The abundance of fissile 2j5U in uranium fuel is an important parameter. The measurement of `U / U ratio in a sample of a nuclear fuel pellet has been an important topic of the nuclear engineering field. This ratio is needed for the production of fuel elements, quality control, calculation of burn - up periods and the control ' of reprocessing of fuel elements. Uranium isotopic analysis is also necessary in some medical studies and environmental samples. The main objective of this study is to determine the isotopic ratio of 2j5U / z38U in uranium samples at various grades of enrichment, by using the instrumental neutron activation analysis technique. Prepared samples are irradiated with thermal and epithermal neutrons and isotopic ratio is calculated from the gamma-ray spectra of irradiated samples. Mass spectrometer gives the isotopic ratio of U / ` U as low as 0.1 %, but such an analysis takes long time and is expensive. Some alternative xvmethods are needed. The less accurate methods which are used for the measurement of U / U ratio can be listed as; mass spectrometer, neutron activation analysis with high resolution gamma spectrometry, passive gamma - ray spectrometry, alpha spectrometry and delayed neutron measurements. A set of experiments have been performed in İstanbul Technical University Institute for Nuclear Energy to measure the 2j5U / 238U ratio. This set includes the experiments of long irradiation time using high resolution gamma ray spectrometry and passive gamma - ray spectrometry. Neutron activation analysis with short irradiation time have been performed at Atom Institute of Austria University. the ratio U / U is determined by the method of neutron activation analysis of long irradiation time. 4.76 % enriched uranium is used to prepare samples. Samples were irradiated at the central channel of the TRIGA Mark II research reactor, which is ran by institute. Gamma - ray of 293 keV of fission product î43Ce is chosen as indicator of 2j5U and gamma - ray of 277 keV of 239Np indicated 238U in the sample. Weight ratio and net peak area ratio is proportional and proportionality constant depends on fisson product gamma ray energy, detector efficiency, neutron energy and experimental conditions. Uranilnitrathexahidrat (unknown enrichment) and 4.76 % enriched uranil- nitrathexahidrat are used to prepare solutions of 15 mg U / ml and 5mg U / ml respectively. Various mixtures of the solutions having different amount of each of them are prepared and dried under the infrared lamp. The samples were irradiated in the reactor running at 250 kW for the duration of 10 minutes. The multichannel spectrometer is calibrated in such a way that 40 channel difference exists between 277 keV and 293 keV (at the range of 4K channel). The results of the experiments were analyzed by the computer code GANAAS. A calibration line which is passing through two points were obtained on the coordinate system (P (fission product) / P (239Np) )versus (W235 / W238) using the above given equation. One point were obtained from the neutron activation analysis results of 4.76 % (235U / 238U atomic ratio = 0.05) enriched uranilnitrathexahidrat and the other was the origin of the coordinate system. The analysis results of the other samples lied on the line and their (W235 / W238) ratio were determined. Unknown enrichment of the uranilnitrathexahidrat were found as 0.003942 ± 0.000179. In fact, this result showed that it is depleted uranium sample. xviAfter this preliminary study, short irradiation time is used for the Tir T To determination of Ç `U / ` U) ratio. Short irradiation time experiments have been performed at Atom Institute of Austria University in Vienna. TRIGA Mark II reactor was used for the irradiation of the samples. Reactor has a fast transfer (150 ms) system from core to the counting set. The possibility of fast transfer and short irradiation time, enabled researcher to count the gamma - rays of j9U (half life 23.45 minute) as the indicator of U in the samples. Gamma - rays of various fission fragments having short half life counted which were produced from the fission of 2j5U nuclei in the samples. Hardening the neutron spectrum in activation analysis causes less matrix activity which gives advantage in evaluating the gamma-ray spectrum. 6LiD converter used to harden the spectrum. Converter placed in and removed from the reactor core with the aid of a pneumatic system. Gamma rays from various fission fragments ( 89Rr, 90Kr, 90Rb, 9SRb, 93Sr, 94Sr, i03Tc, i04Mo, 132Sb, 136I,,37Xe, i39Xe,,40Cs, i43Ba, 144La, i45Ce, i46Ce ) and 2j9U were counted. The ratio of the net peak area is proportional to the ratio of the weights (W235 / W23g)- The lines of peak area ratio versus weight were obtained from the activation and counting of samples of different enrichment. These calibration curves can be used for the determination of the 235U enrichment of the unknown sample. Best result is obtained using 6LiD converter and counting 1428 keV, 397 keV gamma - rays of fission fragments 94Sr and i44La respectively. 75 keV gamma-ray of 239U is counted as an indicator of 2i8U, 6LiD converter is used to harden the spectrum in order -to decrease the effects of the matrix material on the results of the experiments. Samples were irradiated in the reactor for 2 seconds. After a 40s for waiting time, they were counted for 600 s. According to the best of author's knowledge, there is not any study in the literature to determine 235U / 2j8U ratio, counting above mentioned short half life fission fragments and using such a short activation time. Passive gamma-ray spectrometry method is also used for the determination of the same ratio in our laboratory. Passive gamma - ray spectroscopy is based on the counting of 186 keV gamma - rays emitted as XVIIa result of decay of 2j5U and 63 keV gamma-rays emitted from the 234lh which is a decay product of U. The ( `U / U) ratio is obtained using the net peak areas, half life's, detector efficiencies and the probabilities of interested gamma - rays emissions from 2j4Th and 235U. Passive gamma - ray spectroscopy used for the determination of ( `U / *` U) ratio of 4.76 % enriched uranium, natural uranium and depleted uranium. Samples and background were counted for the same duration which is about 24-44 hours. The results of the experiments were analyzed by the computer code GANAAS. The ` U / ` U atomic ratio of 4.76 % enriched uranium which is 0.05, was found as 0.05444 ± 0.00436 and the 235U / 238U atomic ratio of natural uranium, which is 0.007205 was found as 0.00751 ± 0.00059. The same depleted uranium which was used in long irradiation time experiments was analyzed and 235 U / 238U atomic ratio of 0.00405 ± 0.00034 was obtained. Counting system in both laboratories consist of pure germanium detectors, necessary electronics and data analyzing systems. The efficiency of commercial germanium detectors quoted in the list of specification is relative foil - energy peak efficiency with reference to absolute efficiency of 3x3in Nal (Tl) crystal. The measurement is based on the 1.33MeV peak of 60Co. I t is assumed that a 60Co source of known strength is positioned 25cm away from the face of the detector. A count is taken for a period of time, and the absolute full - energy peak efficiency of germanium detector is determined. This absolute efficiency is divided by 1.2 x 10`J,which is the absolute efficiency of a 3x3in Nal(Tl) crystal 25cm from the source, to give the relative efficiency quoted in the specifications. The relative efficiencies of the germanium detectors were checked by this method to distinguish the difference between specified value and recent efficiencies before the analysis. In this study; the known method of neutron activation with long irradiation and waiting time, and passive gamma ray spectrometry are investigated with regards to their use in our laboratory to determine U / U ratio. The result is satisfactory. Besides these methods, a new method based on counting of the characteristic gamma rays originated from L 9U and short half life fission fragments is used for the determination of the same ratio. The advantage of the latter method is its promptness, U / U can be determined in 5 to 20 minutes but other methods mentioned above need 1-3 days at least. XV1UAnalysis were made by using GANAAS (Gamma Activity, Neutron activation Analysis Systems, 1995, IAEA) computer code. Necessary programs were written to transfer the spectrum data from Canberra 90 multichannel! analyzer to the computer and to make the data ready to be used by GANAAS. Thus, a complete spectrum evaluation set is made ready for the evaluation of any gamma - ray spectrum which in the laboratory at our institute. xix